Hi everyone
I am a relatively new user, and that’s why I make some annoying mistakes and often struggle with simple problems.
Is it possible to tally alongside depletion? Like if we have an assembly and we deplete it in 5 steps of 1 day each, as time_steps = [1,1,1,1,1]. So, can we tally any scores fission,flux etc after every 1 day, or after every 1 step?
Actually, I’ve been trying to do so in my code but facing an error; No fission sites banked on MPI rank 0. I have checked the volume of my fuel and other depletable materials and I am quite positive there is no issue with volumes or particles/batch.
Yes, if you create and export tallies to an XML file before running depletion, they will be tallied at each depletion timestep and the results will show up in the openmc_simulation_n#.h5 files (which are just statepoint files that can be loaded with openmc.StatePoint).
Regarding the “No fission sites banked” message, this means that particles are not reaching the fissionable material in your model. If you are able to share your model, someone here can try to debug further.
thank you so much for your response. I didn’t know I was supposed to use openmc_simulation_n#.h5 files for retrieving tally results. The issue has been resolved.